ML20214F534
| ML20214F534 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 11/13/1986 |
| From: | Joshua Wilson Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20214F538 | List: |
| References | |
| NUDOCS 8611250341 | |
| Download: ML20214F534 (16) | |
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UNITED STATES Y'
{gg NUCLEAR REGULATORY COMMISSION 7.
p WASHINGTON. D. C. 20555
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LOUISIANA POWER AND LIGHT COMPANY DOCKET N0. 50-382 WATERFORD STEAM ELECTRIC STATION, UNIT 3 l
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 8 License No. NPF-38 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment, dated May 23, 1986, as supplemented by letters dated August 29, 1986 and October 1, 1986, by Louisiana Power and Light Company (licensee), complies with standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Ae..coingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-38 is hereby amended to read as follows:
8611250341 861113 PDR ADOCK 05000382 P
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. (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 8, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in this license.
LP&L shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION A
J es H. Wilson, Project Manager PWR Project Directorate No. 7 Division of PWR Licensing-B
Attachment:
Changes to thc Technical Specifications Date of Issuance:
November 13, 1986
November 13, 1986
. ATTACHMENT TO LICENSE AMENDMENT NO. 8 TO FACILITY OPERATING LICENSE NO. NPF-38 DOCKET NO. 50-382 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Also to be replaced are the following overleaf pages to the amended pages.
Amendment Paces Overleaf Pages 3/4 4-24 3/4 4-23 3/4 4-25 3/4 4-26 3/4 5-5 3/4 5-6
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3/4 7-31 3/4 7-32 B 3/4 4-5 B 3/4 4-6 6-17 6-17a Page 6-18 is reissued without change.
TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS SAMPLE AND PARAMETER ANALYSIS FREQUENCY DISSOLVED OXYGEN At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CHLORIDE At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> FLUORIDE At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
- Not required with T,yg less than or equal to 250*F i
I WATERFORD - UNIT 3 3/4 4-23
REACTOR COOLANT SYSTEM 3/4.4.7 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.7 The specific activity of the primary coolant shall be limited to:
a.
Less than or equal to 1.0 microcurie / gram DOSE EQUIVALENT I-131, and b.
Less than or equal to 100/E microcuries/ gram.
APPLICABILITY:
MODES 1, 2, 3, 4, and 5.
ACTION:
MODES 1, 2, and 3*:
a.
With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
avg b.
With the specific activity of the primary coolant greater than 100/E microcuries/ gram, be in at least HOT STANDBY with T less than 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
avg A
With T greater than or equal to 500 F.
avg WATERFORD - UNIT 3 3/4 4-24 AMENONENT N0. 8
REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)
ACTION:
(Continued)
MODES 1, 2, 3, 4, and 5:
c.
With the specific activity of the primary coolant greater than 1.0 microcurie /gran DOSE EQUIVALENT I-131 or greater than 100/E microcuries/ gram, perform the sampling and analysis require-ments of item 4 a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.
SURVEILLANCE REQUIREMENTS 4.4.7 The specific activity of the primary coolant shall be determinea to be within the limits by performance of the sampling and analysis ' program of Table 4.4-4.
1 WATERFORD - UNIT 3 3/4 4-25 AMEN 0 MENT N0. 8
TABLE 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMDLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WHICH SAMPLE AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED 1.
Gross Activity Determination At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 1, 2, 3, 4 2.
Isotopic Analysis for DOSE 1 per 14 days 1
EQUIVALENT I-131 Concentration 3.
Radiochemical for E Determination 1 per 6 months
- 1 4.
Isotopic Analysis for Iodine a)
Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 1#, 2#, 3#, 4#, 5#
Including I-131, I-133, and I-135 whenever the specific activity exceeds g;
1.0 pCi/ gram, DOSE S>
EQUIVALENT I-131
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or 100/E pCi/ gram, and M
b)
One sample between 1, 2, 3 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a THERMAL POWER change exceeding 15 % of the RATED THERMAL POWER within a 1-hour period.
- Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
- Until the specific activity of the primary coolant system is restored within its limits.
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' EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.
A visual inspection of the safety injection system sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.
3.
Verifying that a minimum total of 97.5 cubic feet of solid trisodium phosphate dodecahydrate (TSP) is contained within the TSP storage baskets.
4.
Verifying that when a representative sample of 4 i 0.01 grams of TSP from a TSP storage basket is submerged, without agitation, in 4 1 0.1 liters of 120 + 10 *F water borated within RWSP boron concentration limits, the pH of the mixed solution is
,I raised to greater than or equal to 7 within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
I e.
At least once per 18 months, during shutdown, by:
1.
Verifying that each automatic valve in the flow path actuates to its correct position on SIAS and RAS test signals.
2.
Verifying that each of the following pumps start automatically upon receipt of a safety injection actuation test signal:
a.
High pressure safety injection pump.
b.
Low pressure safety injection pump.
3.
Verifying that on a recirculation actuation test signal, the low pressure safety injection pumps stop, the safety injection system sump isolation valves open.
f.
By verifying that each of the following pumps required to be OPERABLE performs as indicated on recirculation flow when tested pursuant to Specification 4.0.5:
1.
High pressure safety injection pumps develop a total head of greater than or equal to 1400 psid for cump A, 1431 psid for pump B and 1429 psid for pump A/B.
2.
Low pressure safety injection pump discharge pressure greater than or equal to 177 psig.
WATERFORD - UNIT 3 3/4 5-5 AMENDMENT NO. 8
EMERGENCY CORE COOLING SYSTEMS
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SURVEILLANCE REQUIREMENTS (Continued) g.
By verifying the correct position of 4..ch electrical and/or mechanical position stop for the following ECCS throttle valves by verifying that each ECCS throttle valve opens to the proper throttled position each time the valve is cycled:
HPSI System LPSI System Valve Number Valve Number a.
SI-225A e.
SI-227A a.
SI-138A b.
SI-2258 f.
SI-2278 b.
SI-1388 c.
SI-226A g.
SI-228A c.
SI-139A d.
SI-226B h.
SI-2288 d.
SI-1398 h.
By performing a flow balance test, during shutdown, following completion of mcdifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying the following flow characteristics:
HPSI System - Single Pump (Cold leg injection mode)
The sum of the injection lines flow rates, excluding the highest flow J
l rate, is greater than or equal to 658 gpm for HPSI Pump A running, l
665 gpm for HPSI Pump B running, and 650 gpm for HPSI Pump A/B running, l
with a maximum differential pressure of less than or equal to 528 psid I
for HPSI Pump A, 472 psid for HPSI Pump B, and 489 psid for HPSI Pump A/B.
HPSI SYSTEM - Single Pump (Hot / cold leg injection mode)
With the system operating in the hot / cold leg injection mode, the hot leg flow must be greater than or equal to 436 gpm and within i 10% of the cold leg flow.
LP5I System - Single Pump Flow for each pump is greater than or equal to 4810 with the total developed head greater than or equal to 268 feet but less than or equal to 292 feet.
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' PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) 4.7.10.1.3 Each fire pump diesel starting 12-volt battery bank and charger shall be demonstrated OPERABLE:
4.
At least once per 7 days by verifying that:
1.
The electrolyte level of each battery is above the plates, and 2.
The overall battery voltage is greater than or equal to 12 volts.
b.
At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery.
c.
At least once per 18 months by verifying that:
1.
The batteries and battery racks show no visual indication of physical damage or abnormal deterioration, and 2.
The battery-to-battery and terminal connections are clean, tight, free of corrosion, and coated with anticorrosion material.
WATERFORD - UNIT 3 3/4 7-31 AMENDMENT NO. 8
4 PLANT SYSTEMS SPRAY AND/0R SPRINKLER SYSTEMS
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L_IMITING CONDITION FOR OPERATION 3:7.10.2 The following spray and/or sprinkler systems shall be OPERABLE:
Sprinkler No.
Bldg./Elev.
Location FPM-1 RCB Reactor Coolant Pumps 1A, 18 FPM-2 RCB Reactor Coolant Pump 2A, 28 FPM-3A RAB +21, +46 Diesel Generator Area A, Feed Tank Room A FPM-48 RAB +21, +46 Diesel Generator Area B, feed Tank Room B FPM-11A RAB -35 Emergency D/G Fuel Oil Tank A FPM-128 RAB -35 Emergency D/G Fuel Oil Tank B FPM-16 FWPH +15 Fire Water Pump House FPM-17 RAB +35 Cable Vault Area FPM-18 RAB +35 Electrical Penetration Area 1 FPM-19 RAB +35 Electrical Penetration Area 2 FPM-22 RAB -4 Corridor and 810wdown Tank Rooms FPM-23 RAB -35 Corridor, Shutdown Heat Exchanger Rooms, EFW Pump Room s
FPM-24 RAB +21 Corridors, CCW Area FPM-258 RAB +21 North High Voltage Switchgear Room FPM-26 RAB +46 Ventilation Equipment Rooms FPM-27 RAB +7 HVAC Rooms FPM-28 RAB -35 Auxiliary Component Cooling Water Pump Rooms FPM-29 RAB +35 Relay Room, Corridor FPM-30A RAB +21 South High Voltage Switchgear Room APPLICABILITY:
Whenever equipment protected by the spray / sprinkler system is required to be OPERABLE.
ACTION:
a.
With one or more of the above required spray and/or sprinkler systems inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged unless the spray and/or sprinkler system (s) is located inside the containment, then inspect that containment area at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or monitor air temperature at least once per hour at the locations listed in Specifi-cation 4.6.1.5; for other areas, establish an hourly fire watch patrol.
b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
m.
BASES CHEMISTRY (Continued) the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System i
over the life of the plant.
The associated effects of exceeding the oxygan, chloride and fluoride limits are time and temperature dependent.
Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.
The time interval.
permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.
The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4.7 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2-hour doses at the site boundary will not exceed an appro-priately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady-state primary-to-secondary steam generator leakage rate of 1 gpm and a concurrent loss-of offsite electrical power.
The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations.
These values are conservative in that specific site parameters of the Waterford Unit 3 site, such as site boundary location and meteorological conditions, were not considered in this evaluation.
WATERFORD - UNIT 3 8 3/4 4-5 AMENDMENT NO. 8
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BASES SPECIFIC ACTIVITY (Continued)
Reducing T,yg tolessthan500*Fpreventsthereleaseofactivitysbould a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.
The surveillance requirements provide adequate assurance that excessive i
specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
Information obtained on iodine spiking will j
be used to assess the parameters associated with spiking phenomena.
A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
3/4.4.8 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand I
the effects of cyclic loads due to system temperature and pressure changes.
i These cyclic loads are introduced by normal load transients, reactor trips, i
and startup and shutdown operations.
The various categories of load cycles used for design purposes are provided in Section 3.9.1.1 of the FSAR.
During l
startup and shutdown, the rates of temperature and pressure changes are limited
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so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.
These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.
Therefore, a pressure-temperature curve based on steady-state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.
The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location.
The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel.
These stresses are additive to the pressure induced tensile stresses which are already present. The l
thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.
Consequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.
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MAVS@@A@n - INT T 1 Q 1/44-6
ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code l
of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the hRC unless otherwise noted.
STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.
6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.
Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.
6.9.1.3 Startup reports shall'be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.
If the startup report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation), supplementary reports shall be submitted at least every 3 months.until all three events have been completed.
ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.
The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.7 shall be submitted annually in accordance with the aforementioned. time frame.
The following information shall be included:
(1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded
- This tabulation supplements the requirements of $20.407 of 10 CFR Part 20.
WATERFORD - UNIT 3 6-17 AMENDMENT NO. 8
ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued) and results of one analysis after the radiciodine activity was reduced to less than limit.
Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.
6.9.1.5 Reports required on an annual basis shall include a tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions * (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance
[ describe maintenance], waste processing, and refueling).
The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements.
Small exposures totalling less than 20% of the indi-vidual total dose need not be accounted for.
In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions.
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l WATERFORD - UNIT 3 6-17a AMENDMENT NO. 8
ADMINISTRATIVE CONTROLS
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MONTHLY OPERATING REPORTS 6.9.1.6 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.
ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6.9.1.7 Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitt4d prior to May 1 of each year.
The initial report shall be submitted prior to May 1 of the year following initial criticality.
.The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, with operational controls as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment.
The reports shall also include the results of land use censuses required by Specification 3.12.2.
The Annual Radiological Environmental Operating Reports shall include the j
results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the lo'ations specified in the Table and Figures in the 00CM, as well as summarized c
and tabulated res'ults of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979.
In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.
The missing data shall be submitted as soon as possible in a supplementary report.
The reports shall also include the following:
a summary description of the radiological environmental monitoring program; at least two legible maps
- covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 3.12.3; discussion of all deviations from the sampling schedule of Table 3.12-1; and discussion of all analyses.in which the LLO required by Table 4.12-1 was not achievable.
- 0ne map shall cover stations near the SITE BOUNDARY a second shall include the more distant stations.
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